ML20081F906
| ML20081F906 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/17/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20081F902 | List: |
| References | |
| NUDOCS 9503220327 | |
| Download: ML20081F906 (10) | |
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q 4-UNITED STATES 1
E NUCLEAR REGULATORY COMMISSION.
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WASHINGTON, D.C. 2006 Hoot SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 130 TO FACILITY OPERATING LICENSE NPF-35 i
AND AMENDMENT NO.124 T0' FACILITY OPERATING LICENSE NPF-52 '
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- CATAWBA NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 i
1.0 INTRODUCTION
t By letter dated November 29, 1994, as supplemented by letters dated January 12, January 27, February 6, February 16, and March 13, 1995, Duke Power Company, et al. (the licensee), submitted a request for changes to the.
Catawba Nuclear Station, Units 1 and 2, Technical Specifications (TS). The requested amendments ~ revise, in part, TSs 4.4.5.2, 4.4.5.4, 4.4.5.5 and 3.4.8) for Catawba Unit 1, Cycle 9 operation to permit the use of a voltage-based.
-steam generator tube repair criteria for defects confined within the thickness of the tube support plate. The February 6 and 16, and March 13, 1995, letters provided clarifying information that did not change the scope of the November 29, 1994, application and the initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
The-staff has previously approved similar requests from the licensee to apply the voltage-based tube repair criteria at Catawba.
Implementation of the voltage-based tube repair criteria for the seventh operating cycle was approved as documented in an amendment to the license dated September 25, 1992, " Issuance of Amendments - Catawba Nuclear Station, Unit 1 (TAC No.
M84221)," and in a letter dated July 30,1993, " Catawba Nuclear Station, l
Unit.1 - Safety Evaluation Regarding Steam Generator Tube Interim Plugging Criteria Mid-Cycle Inspection (TAC No. M86116)" (referred to as Reference I and Reference 2, respectively). Similarly, implementation of the voltage-based tube repair criteria for the eighth operating cycle was approved t-as documented in an amendment to-the license dated December 16, 1993,
" Issuance of Amendments - Catawba Nuclear Station, Units 1 and 2 Steam Generator Interim Plugging Criteria for Unit 1, Cycle 8 (TAC Nos._ M87840 and M87841) (Reference 3)."
In References 1, 2, and 3, the staff concluded that the tube repair limits and leakage limits would ensure adequate structural and leakage integrity of the steam generator tubing at Catawba Nuclear Station, Unit 1, consistent with applicable regulatory requirements, for the seventh and eighth operating cycles. This evaluation addresses comparable tube repair i
limits for operating Cycle 9.
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The staff is currently developing a generic interim position on voltage-based limits for outside diameter stress corrosion cracking (00 SCC) confined to within the thickness of the tube support plates.
The staff has published several conclusions regarding voltage-based repair criteria in draft NUREG-1477, " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes" and in a draft Generic Letter titled " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes." The latter document was published for public comment in the Federal Reaister on August 12, 1994 (FR 59 41520).
However, the staff is continuing to evaluate an acceptable generic position which will take into consideration public comments on the draft Generic Letter cited above, domestic operating experience under the voltage-based repair criteria, and additional data which have been made available from European nuclear power plants. The staff currently plans to document its final position on this matter in a Generic Letter. Pending completion and issuance of the staff's final generic position on the voltage-based tube repair criteria, the staff is continuing to evaluate voltage-based repair criteria proposals on a case-specific basis.
Each of these case-specific evaluations of the voltage-based repair criteria is limited to one cycle of operation.
The licensee's current proposal is applicable to Cycle 9 operation and is similar to the licensee's previous proposals which were approved as documented in References 1, 2, and 3.
Furthermore, the licensee's submittal is i
consistent with the draft Generic Letter issued for public comment on August 12, 1994, except as noted below.
3.0 PROPOSED INTERIM TUBE REPAIR CRITERIA Catawba Nuclear Station, Unit 1, TSs 4.4.5.2, 4.4.5.4, and 4.4.5.5 are revised by this amendment request to specify the voltage-based tube repair criteria for ODSCC confined to within the thickness of the tube support plate. A few i
modifications were made to the previously approved technical specifications pertaining to the implementation of the voltage-based tube repair criteria.
The changes in the technical specifications for Cycle 9 implementation of the voltage-based tube repair criteria include, in part:
a.
Specifying that the determination of the cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least 20 percent random sampling of tubes inspected over their full length.
(November 29, 1994, letter) b.
Eliminating an outdated reference for the calculation of postulated main steam line break (MSLB) primary-to-secondary leakage.
Removing a sentence that specified that no RPC inspections were required for tubes that will be administrative 1y plugged or repaired.
Providing an equation that specifies the appropriate tube repair limits for unscheduled mid-cycle inspections not attributable to leakage from 00 SCC at the support plate intersections.
t i' e.
Replacing the previous reporting requirements with the following:
l (November 29, 1994, letter) e For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service.should any of.the following conditions arise:
1..
If th'e estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing licensing basis assumptions) during the previous operating cycle.
i 2.
If circumferential crack-like indications are detected at the tube support plate intersections.
3.
If the indications are identified that extend beyond the l
confines of the tube support plate.
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1 x 10', calculated conditional burst probability exceeds -
If the 4.
, notify the NRC and provide an assessment of the i
safety significance of the occurrence.
In addition tr the above technical specification changes, the licensee also:
made the following commitments for implementing the. voltage-based repair criteria:
1.
It will follow the guidance of the draft Generic Letter with 3 exceptions. The exceptions are in the areas of bobbin coil calibration, periodic tube pulling, and probe wear-standard.
(November 29, 1994, letter).These exceptions are discussed in Sections 4.1, 4.2.1 and 4.1, respectively, of this report.
2.
NRC-accepted methodologies for calculation of the conditional probability of burst and total leak rate during a MSLB will be utilized.
3.
The NRC will be notified prior to restart if any indications of primary water stress corrosion cracking (PWSCC) are detected at the tube support plate elevations.- Furthermore, the data analysts will be briefed on the possibility that PWSCC can occur at tube support plate elevations.
4.
A tube pull will be-performed in the event of a forced outage on' Unit I during Cycle 9 if the forced outage is due to a steam generator tube leak attributable to 00 SCC at a tube support plate-intersection.
All known tube leaks will be repaired.
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The repair criteria will not be applied to the flow distribution baffle plate intersections.
(February 6,1995, letter) i 7.
The amendment will only be applicable to Cycle 9 operation.
(January 12, 1995, letter) 4.0 EVALUATION 4.1 Insnection Issues The licensee intends to incorporate the inspection guidance of the draft Generic Letter into their inspection program with the exception of the bobbin coil calibration standard and the probe wear re-inspection requirements. The i
licensee also intends to use a 0.630-inch diameter bobbin probe in lieu of a 0.610-inch diameter bobbin probe.
For the bobbin coil calibration, the licensee intends to calibrate the bobbin coil on the 4-20% holes rather than the 4-100% holes recommended in the draft Generic Letter.
Furthermore, the licensee proposes not to use a wear standard for the general inspection but rather to reinspect only those indications with voltages greater than 0.7 volts with a probe controlled to the wear standard variability limit.
The licensee has calibrated the bobbin coil on the 4-20% through-wall holes,,
since initial implementation of the voltage-based tube repair criteria in 1992. The staff has concluded that calibrating on the 4-20% through-wall holes rather than the 4-100% through-wall holes is acceptable based on (1) a review of the material provided in EPRI report NP-7480-L Volume 1 pertaining to assessing the use of 20% and 100% through-wall holes and 100% through-wall EDM slots, and (2) the results obtained by an independent contractor j
pertaining to the repeatability of voltage measurements between standards containing 20% through-wall holes,100% through-wall holes, and 100% EDM l
notches. These two studies showed that the 20% through-wall holes were more i
reproducible and the voltage readings obtained on these holes were more i
repeatable. Although deeper defects are typically the ones of most concern I
and the 100% through-wall holes are more representative of these defects, the staff has concluded that the better reproducibility of the 20% holes and the
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better measurement repeatability provided with these holes, in conjunction j
with the limits on new probe variability on the other holes in the standard (i.e., the 40%, 60V, 80%, an.1 100% through-wall holes), justifies calibrating the bobbin coil on the 4-20% through-wall holes.
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With' respect to the use of an alternate procedure for re-inspecting tubes that fail to meet the probe wear criterion and the use of a 0.630-inch diameter bobbin coil, the staff has concluded that alternate methods.such as these may be used generically and on a continuing basis'provided an assessment is performed demonstrating that (1) they provide equivalent detection. and sizing capability on a statistically significant basis when compared to tne guidance in the draft Generic Letter, and (2) they are consistent with current methods for determining the end-of-cycle (E0C) voltage distributions which are used in the tube integrity analyses. These assessments, along with the statistical criteria for demonstrating that the techniques are equivalent, should be provided to the NRC for review and approval. With respect to this cycle-specific application, however, the staff has concluded that the methods which
- have been previously employed for reinspecting tubes when a probe fails to meet the probe wear criterion are acceptable. The use of the 0.630-inch bobbin coil is discussed in the following paragraphs.
By letter dated March 13, 1995, the licensee provided the preliminary results of a study comparing the voltage response from the 0.630-inch and 0.610-inch diameter bobbin coil probes. The staff has concluded that using the 0.630-inch diameter probe is acceptable for this cycle of operation at Catawba Unit 1 provided that the bobbin voltages for indications measured with the 0.630-inch diameter probe are scaled upward by a factor of 1.25 to determine a voltage which would be comparable to that which would have been obtained if a 0.610-inch diameter probe had been used. The 0.610-inch diameter bobbin probe is the probe that was used in the development of the voltage-based repair criteria. As a result of implementing this criteria for measurements made with a 0.630-inch diameter bobbin probe, (1) bobbin indications less than or equal to 0.8 volts (0.630-inch probe) can remain in service, (2) bobbin indications in excess of 0.8 volts (0.630-inch probe) and less than 2.2 volts (0.630-inch probe) can be left in service provided an RPC inspection of the indication does not detect ODSCC or any other degradation mode, and (3) crack indications above 2.2 volts (0.630-inch probe) must be repaired regardless of whether the indication was confirmed with an RPC probe.
In addition to the above criteria for the 0.630-inch diameter bobbin probe, the licensee may still implement the 1.0/2.7 volt criteria (versus the 0.8/2.2 volt criteria)/
when the 0.610-inch diameter bobbin probe is used for the inspection.
The licensee believes that the 1.25 adjustment factor for converting the 0.630-inch diameter probe readings to an equivalent 0.610-inch diameter probe reading is conservative; however, in the March 13, 1995, letter, the licensee also committed to perform a more detailed assessment of the data and to assess the non-destructive examination error associated with the use of the 0.630-inch diameter probe. The staff will review this material to determine the generic applicability of using the 0.630-inch probe and its impact on the current tube integrity analyses (i.e., the conditional probability of burst and the conditional leak rate calculations) which are discussed below.
As a result of the potential for the possible development of primary water stress corrosion cracking (PWSCC) flaws at dented tube support plate intersections, the licensee has briefed their eddy current analysts of the potential for PWSCC to occur at these locations.
Furthermore, the licensee has agreed to notify-the NRC prior to plant restart if any PWSCC indications are detected at the support plate elevations. The staff notes that if PWSCC is detected at support plate elevations, an evaluation to ensure the voltage-based repair criteria is only applied to 00 SCC indications may need to be performed.
In summary, the staff concludes that the inspection guidelines submitted by the licensee are acceptable since the proposed repair criteria is limited to one cycle, and the calibration, recording, and analysis requirements are consistent with the methodology used in the development of the databases and supporting evaluations.
. 4.2 Structural IntearitY 4.2.1 Deterministic Structural Intearity /;
sment The licensee's tube repair limits are based on a correlation between the burst pressure and the bobbin voltage of pulled tube and model boiler data. This correlation is similar to that used in approving the voltage limits in the licensee's previous submittals and those used in the draft Generic Letter.
The staff finds the licensee's voltage limits acceptable given the current burst pressure / bobbin voltage database, the licensee's projected growth rates, and the non-destructive examination uncertainty estimates.
To confirm the nature of the degradation occurring at the tube support plate elevations, tubes are periodically removed from the steam generators for destructive analysis. Tube pulls confirm that the nature of the degradation being observed at the tube support plate elevations is predominantly axially oriented ODSCC and also provide data for assessing the reliability of the inspection methods and for supplementing existing databases (e.g., burst pressure, probability of leakage, and leak rate). The draft Generic Letter contains guidance that states utilities should remove 6 intersections for destructive examination every other outage.
To follow the draft Generic Letter guidance on tube pulls, the licensee would need to pull 6 intersections from their steam generators during this outage since their last tube pulls were two outages ago.
Since 1990, the licensee has removed 9 tubes representing 18 intersections from the Catawba Unit I steam generators. The examinations performed on these pulled tubes confirmed that the dominant degradation mechanism for the indications at the support plate elevations was axially oriented ODSCC and that the indications at these intersections are consistent with the data in the databases accumulated from other plants with ODSCC at tube support plate intersections.
The voltage-based tube repair criteria are part of the licensee's short-term strategy for addressing steam generator tube degradation. The licensee's long-term strategy involves replacing the steam generators at the end of their upcoming cycle (i.e.,
Cycle 9). The staff has concluded that the licensee does not need to pull tubes during the upcoming refueling outage since the licensee intends to replace their steam generators at the end of the forthcoming cycle (i.e.,
Cycle 9). However, the licensee has committed to perform a tube pull in the event of a forced outage on Unit I during Cycle 9, if the forced outage is due to a steam generator tube leak as a result of ODSCC at a tube support plate intersection.
4.2.2 Probabilistic Structural Intearity Assessment A probabilistic analysis for the potential for steam generator tube ruptures, given a Main Steam Line Break (MSLB), has been performed for the previous applications of this tube repair criteria.
The draft Generic letter contains additional guidance on this analysis. The licensee intends to perform this calculation per the guidance in the draft Generic Letter which will most likely result in a higher conditional probability of burst than would have been obtained using the previous methodology since it includes parametric uncertainty. The results of the probabilistic analysis will be compared to a threshold value of I x 10-2 per the guidance in the draft Generic Letter.
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. This threshold value will provide assurance that the probability of burst is acceptable considering the assumptions of the calculation and the results of the staff's generic risk assessment for steam generators contained in i
NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." Failure i
to meet the threshold value indicates that ODSCC confined to within the thickness of the tube support plate could contribute a significant fraction to the overall conditional probability of tube rupture from all forms of degradation that was assumed and evaluated as acceptable in NUREG-0844.
The licensee supplied the details of the methodology for calculating the conditional probability of burst given a MSLB by letter dated February 16, 1995. The staff finds the licensee's proposal to perform the calculation per the guidance in the draft Generic Letter to be acceptable for this outage-specific application. As noted above, the staff expects this calculation to result in a higher probability of burst than would have been calculated previously since it includes parametric uncertainty. The staff notes that all applicable data should be included in the burst pressure database when performing this calculation, except as discussed below.
4.2.3 Data Exclusion from the Burst Pressure Database During the performance of the pulled tube examinations, malfunctions in the test equipment or improper specimen preparation can occasionally occur which could result in erroneous readings.
Data such as this should not be included in a database since it could result in invalid results and/or conclusions.
The staff, therefore, concluded in draft NUREG-1477 that eliminating data from the burst pressure database was appropriate provided that the data could be shown to be erroneous or the result of an invalid test.
The staff provided additional guidance regarding the exclusion of data from the burst pressure database in a meeting with the industry on February 8, 1994. As a result of this guidance, the industry provided criteria (i.e., data exclusion criteria) for determining whether data may be removed from the burst pressure database in a letter dated June 9,1994, from David J. Modeen of the Nuclear Energy Institute (NEI) to Dr. Brian Sheron of the NRC. This letter was referenced and discussed in the draft Generic Letter. The licensee's letter of November 29, 1994, addressed its plans to follow the guidance of the draft Generic Letter which included the issue of data exclusion.
The staff has concluded that the exclusion of the burst pressure data points cited in the June 9,1994, letter, from the burst pressure database is appropriate. However, pending further review of the generic data exclusion criteria presented in the June 9, 1994, letter, the staff is continuing to assess the appropriateness of excluding data points from the burst pressure database on a case-by-case basis.
4.3 Leakaae Intearity 4.3.1 Normal Operational Leakagg Consistent with prior amendments approving the use of the voltage-based repair criteria at Catawba Unit 1, the licensee will continue to limit the amount of
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operating leakage through any one steam generator to 150 gallons per day (gpd). This requirement is currently in the technical specifications for Catawba Unit 1.
4.3.2 Accident Leakaae I
The licensee has proposed a model for calculating the steam generator tube i
leakage from the faulted steam generator during a postulated MSLB which consists of two major components:
(1) a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability.
of leakage (POL) model); and (2) a model predicting leak rate as a function of voltage, given.that leakage occurs (i.e., the conditional leak rate model).
t The calculational methodology being proposed by the licensee for Catawba, Unit 1 for determining the amount of primary-to-secondary leakage under postulated accident conditions has previously been reviewed and approved by the staff as documented in the Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 54 to Facility Operating License NPF-72, Commonwealth Edison Company, Braidwooo Station, Unit 1, Docket No. STN 50-456 dated August 18, 1994. The staff finds this methodology acceptable for Catawba Unit 1.
The staff notes that all applicable data should be included in the probability of leakage and conditional leak rate databases when performing this calculation, except as discussed below.
The licensee has calculated the allowable steam generator leak rate to be 17.5 gallons per minute (gpm) in the faulted steam generator. This value is intended to be consistent with maintaining the radiological consequences of a release outside containment to within a small fraction of the guideline values in 10 CFR Part 100.
As a result, if the primary-to-secondary leakage during a postulated MSLB is less than the 17.5 gpm limit, the steam generator tubing will maintain adequate leakage integrity under these conditions.
4.3.3.
Data Exclusion from the Leakaae Databases During the performance of the pulled tube examinations, malfunctions in the test equipment or improper specimen preparation can occasionally occur which could result in erroneous readings. Data such as this should not be included in a database since it could result in invalid results and/or conclusions.
The staff, therefore, concluded in draft NUREG-1477 that eliminating data from the steam generator 1eakage databases (i.e., the probability of leakage and the conditional leak rate databases) was appropriate provided that.the data could be shown to be erroneous or the result of an invalid test.
The staff provided additional guidance regarding the exclusion of data from the steam generator leakage databases in a meeting with the industry on February 8, 1994. As a result of this guidance, the industry provided criteria (i.e.,
data exclusion criteria) for determining whether data may be removed from the leakage databases in a letter dated June 9, 1994, from David J. Modeen of'the Nuclear Energy Institute (NEI) to Dr. Brian Sheron of the NRC. This letter was referenced and discussed in the draft Generic Letter.
The licensee's letter of November 29, 1994, addressed its plans to follow the guidance of the draft Generic Letter which included the issue of data exclusion.
-g-The staff has concluded that the exclusion of the probability of leakage data points cited in the June 9,1994, letter, from the probability of leakage database is appropriate.
Furthermore, the staff has concluded that exclusion of the conditional leak rate data points cited in the June 9, 1994, letter, from the conditional leak rate database, with the exception of model boiler specimen 598-3, is appropriate. However, pending further review of the generic data exclusion criteria presented it the June 9, 1994, letter, the staff is continuing to assess the appropriateness of excluding data points from the leakage databases on a case-by-case basis.
In the June 9, 1994, letter, an analysis of the leak rate expected from an indication on tube R28C41 removed from another nuclear power plant (identified as Plant S) was performed. The leakage from this specimen exceeded the capacity o' t.he test facility and, as a result, an assessment of the leakage from this specuen was performed.
Pending further staff evaluation of the appropriate leakage value for this data point, the staff has concluded that this data point thould be assigned a leakage value of 2496 liters per hour (1/hr) consistent with the leakage predicted using the CRACKFLO model.
5.0
SUMMARY
OF EVALUATION Based on the above evaluation, the staff concludes that adequate structural and leakage integrity can be ensured, consistent with applicable regulatory requirements, for indications to which the voltage-based repair criteria will be applied during Cycle 9 at Catawba Nuclear Station Unit 1.
The staff's approval of the proposed voltage-based repair criteria is based, in part, on the licensee being able to demonstrate that the conditional probability of burst and the primary-to-secondary leakage during a postulated MSLB will be acceptable.
6.0 DOSE ASSESSMENT The licensee proposed to amend TS 4.4.5.4.a.13)1 to decrease the assumed primary-to-secondary leakage associated with a postulated MSLB from 30 gallons per minute (gpm) to 17.5 gpm. The licensee also proposed to return the allowable value of the primary coolant activity. in TS 3.4.8 for Unit 1 to 1.0 micro curie per gram from its previous value of 0.58 micro curie per gram.
The licensee presented the results of its analysis of the dose consequences of a MSLB using these parameters to demonstrate the acceptability of the proposed changes.
The staff has independently calculated the doses resulting from a MSLB.
It should be noted that the staff performed its calculation in accordance with the methodology associated with SRP 15.1.5, Appendix A and that the staff did not credit the analysis with the removal associated with the letdown flow and the removal of radioiodine by the letdown demineralizer upon initiation of the accident.
The latter assumptf ?' was made by the licensee in its calculations.
However, the staff concluded thn assuming credit for such removal is not appropriate because the letdown demineralizers are never tested to demonstrate iodine removal capability (as Engineered Safety Feature adsorbers and HEPA filters are) and the letdown demineralizers are not safety-related components.
The results of the staff's calculations confirm the licensee's conclusions
. that the doses would be within the limits established by the Standard Review Plan (SRP) 15.1.5, Appendix A and are acceptab').
7.0 STATE CONSULTATION
In accordance with the Comission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no coments.
8.0 ENVIRONMENTAi.CONSIDERATIQM The amendments change requirements with respect to installation or use of a facility component located within the restricted area a:, defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and thera has been no public coment on such finding (60 FR 7801 dated February 9, 1995). Accordingly, the amendments meet the f
eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement-or environmental assessment need be prepared in connection with the issuance of the amendments.
9.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendments will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributors:
K. Karwoski J. Hayes Date:
E rch 17, 1,995